The present invention generally relates to a testing method for nuclear fuel materials, and more particular to a testing method for multi-component nuclear fuel particles which is characterized by a high level of accuracy and efficiency.
Recent developments in nuclear reactor technology have created a corresponding need for improved fuel materials having a high level of structural integrity. In particular (as discussed in further detail below), fuel particles have been developed for high temperature gas reactor systems (hereinafter known as "HTGR" systems) which involve small, substantially spherical particles (microspheres) having an average diameter of about 300-900 .mu.m. Each of these particles includes a central core or center portion comprised of a fissionable radioactive material. In a preferred embodiment, this material will consist of .sup.235 UCO (uranium-235 carbonate). The center portion or central core of each particle is entirely covered/encapsulated by multiple protective layers preferably consisting of pyrolytic carbon, as well as at least one barrier layer preferably consisting of silicon carbide (SIC). The pyrolytic carbon layers are optimally applied by chemical vapor deposition using a conventional fluidized bed system. The SiC layer is preferably derived by the thermal decomposition of methyltrichlorosilane. The completed particles which incorporate the foregoing chemical compositions are often called "TRISO" particles, and are further discussed in the following references (incorporated herein by reference) which likewise discuss HTGR technology: Tennery, V. J., et al., "Structural Characterization of HTGR Pyrocarbon Fuel Particle Coatings", J. Am. Ceram. Soc., 60(5-6):268-274(1977); Stinton, D. P., et al., "Effect of Deposition Conditions on the Properties of Pyrolytic SiC Coatings for HTGR Fuel Particles", Ceramic Bulletin, 57(6):568-573(1978); Krautwasser, P., et al., "Raman Spectral Characterization of Silicon Carbide Nuclear Fuel Coatings", J. Am. Ceram. Soc., 66(6):424-433(1983); Smith, C. L., "SiC-Fission Product Reactions in HTGR TRISO UC.sub.2 and UC.sub.x O.sub.v Fissile Fuel: I., Kinetics of Reactions in a Thermal Gradient", J. Am. Ceram. Soc., 62(11-12):600-606(1979); and Allen, P. L., et al., "Nuclear Fuel Coated Particle Development in the Reactor Fuel Element Laboratories of the U.K. Atomic Energy Authority", Nucl. Technol., 35:246-253(1977). Furthermore, while the present invention shall be described herein with reference to a nuclear fuel particle containing a .sup.235 UCO center region with multiple pyrolytic carbon protective layers and at least one SiC barrier layer, the present invention may likewise be used in connection with nuclear fuel particles of comparable physical character/dimensions which contain other materials aside from those listed above. Further information regarding the physical, chemical, and structural character of nuclear fuel materials suitable for testing in accordance with the present invention shall be discussed in greater detail below.
Of particular importance regarding the use of nuclear fuel particles (e.g. particles having a radioactive core/center portion surrounded by at least one protective layer and at least one barrier layer) is the physical strength and integrity of each particle with emphasis on the barrier layer. As indicated above, a preferred barrier layer associated with HTGR fuel particles of the type described herein is comprised of SiC. This material is chemically characterized as a moderately brittle ceramic composition. The barrier layer is of particular importance since a significant amount of the strength and structural integrity of each fuel particle is directly attributable to the barrier layer associated therewith. In addition, the barrier layer is designed to retain fission products (e.g. xenon, krypton, carbon monoxide, cerium, cesium, and palladium) within each particle unit during use in an HTGR system. The presence of a weak and ineffective barrier layer in a nuclear fuel particle will diminish the strength/durability of the particle, and will also permit the leakage of fission products outwardly from the particle. For this reason, it is desirable to test the structural integrity of a particle sample before using a particular batch or supply of fuel particles within a selected reactor system. In this regard, the present invention involves a new and unique method for testing nuclear fuel particles as discussed in further detail below.
When nuclear fuel particles and brittle ceramic materials therein (e.g. SiC) are tested for mechanical strength, they exhibit a wide sample-to-sample variation in measured strength values. Strength distribution and stress analysis results are also affected by the selected test method. Many prior testing methods have been used to test the strength and structural integrity of "TRISO"-type nuclear fuel particles. For example, strength tests have been conducted using diametrical compression involving rings of SiC barrier layers removed from TRISO particles containing a center region comprised of .sup.235 UCO as discussed in Bongartz, K., et al., "The Brittle Ring Test: A Method for Measuring Strength and Young's Modulus on Coatings of HTR Fuel Particles", J. Nucl. Mater., 62:123-137(1976). Testing as described in the foregoing article involved the production of ring sections from each test particle using parallel cuts through the particle. Compressive force was thereafter applied to each ring section until it fractured in order to generate information regarding stress characteristics of the ring section under consideration. However, in many instances involving the use of this procedure, structural damage occurred to the ring sections during removal from each particle. Ring sections damaged during processing (e.g. cutting and polishing) were thereafter discarded since they could not be effectively tested. As a result, data was lost for many ring sections, especially those having inherent defects or weaknesses which could have generated valuable comparative information. Furthermore, when an individual ring section is tested using diametrical compression, only a small portion of the inner barrier layer (e.g. SiC) associated with each particle is exposed to maximum tensile compression. A particular ring section may represent only 10% of a particle's SiC surface area. In this regard, the area under maximum tensile stress may be about 10% (or less) of the ring section. For any particle being tested, use of the foregoing test procedure will therefore expose only about 1% of a given particle to maximum stress levels. In contrast, when the selected fuel particles are actually used in a reactor system, the entire surface and volume of the barrier layer (SIC) is exposed to maximum stress levels.
Another testing technique is discussed in Gilchrist, K. E., et al., "A Technique for Measuring the Strength of High Temperature Reactor Fuel Particle Coatings", J. Nucl. Mater., 43:347-350(1972). This technique involved a probability-based method designed to test the surface and interior volume of the barrier layer (SIC) in each particle. To implement this test, various portions of each test particle were physically removed (e.g. by cutting and the like), ultimately resulting in the preparation of a hollow hemispherical section from the particle. The hemispherical section was then cemented over a small hole in a metal (copper) plate and internally pressurized to determine the amount of pressure necessary to fracture the section. Further information regarding this technique is disclosed in Allen, P. L., et al. "Nuclear Fuel Coated Particle Development in the Reactor Fuel Element Laboratories of the U.K. Atomic Energy Authority", Nuclear Technology, 35:246-253(1977).
Finally, an additional method is disclosed in Minkato, K., et al., "Crushing Strength of Irradiated TRISO Coated Fuel Particles", J. Nucl. Mater., 119:326-332(1983). The method disclosed in this reference (hereinafter referred to as the "point load test") involved a crush test designed to determine the strength of selected fuel particles. Specifically, individual particles were positioned between flat platens of hardened steel and compressed between the platens. This method is particularly characterized by a process in which limited portions of the selected fuel particle (e.g. those portions or "points" touching each flat platen) are exposed to stress levels compared with the present invention which more broadly distributes compressive forces. The considerable benefits associated with the broad distribution of compressive forces, as well as further technical and substantive comparisons between both methods will be discussed below.
The present invention involves a unique and highly efficient method which is characterized by numerous benefits compared with prior testing methods including but not limited to: (1) the avoidance of potentially-destructive process steps which involve the physical removal by cutting and the like of various sections of the selected fuel particles; (2) an absence of process steps involving the use of adhesive agents or other materials designed to retain various portions of test particles within the selected testing apparatus; and (3) the use of a process which more broadly distributes compressive forces over test particles, thereby resulting in more accurate, complete, and comprehensive data involving structural integrity, stress capability, and the like. In this regard, the present invention provides numerous advantages compared with prior methods in terms of effectiveness, accuracy, and simplicity. For this reason, the invention described herein represents an advance in the art of nuclear fuel testing as discussed in greater detail below.